IR 05000336/1989024
| ML20012C275 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/08/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20012C272 | List: |
| References | |
| 50-336-89-24, NUDOCS 9003200428 | |
| Download: ML20012C275 (29) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.:
50-336/89-24 Docket No.:
50-336 License No.
DPR-65 Licensee:
Northeast Nuclear Energy Company
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P.O. Box 270 Hartford, Connecticut 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit 2 (
Inspection at: Waterford, Connecticut Inspection Conducted:
December 5, 1989 - January 19, 1990 Reporting Inspector:
Peter J. Habighorst, Resident Inspector Inspectors:
William J. Raymond, Senior Resident Inspector, Millstone Peter J. Habighorst, Resident Inspector, Millstone 2 William Oliveira, Reactor Engineer, Operational Programs Section, Division of Reactor Safety Approved by:
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Donald R. Haverkamp, Chief
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Date Reactor Projects Section 4A Division of Reactor Projects Inspection Summary:
Inspection on December 5, 1989 - January 19, 1990 (Inspection Report No. 50-336/89-24)
' Areas Inspected:
Routine NRC resident inspection of plant operations, radio-logical controls, maintenance / surveillance, engineering / technical support, security, and safety assessment / quality verification including committee activities and Licensee Event Reports. Within certain of these areas, the inspection included review of licensee actions in response to allegations and previously identified items.
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General Conclusions on Adequacy. Strength or Weakness in
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1.icensee Programs Inadequate timeliness of a reportability evaluation and an example of a licensee event report (LER) inaccuracy were noted in review of LER
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89-011-00 and LER 89-009, respectively. (Section 8.3)
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Extensive licensee efforts were not successful in maintaining sufficient i -
reliability of the steam jet air ejector radiation monitor. A significant amount of corrective maintenance was noted without achieving continual
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reliability of the monitor. (Section 4.3.1)
The establishment of a pilot predictive maintenance program for rotational equipment was noteworthy. (Section 5.4)
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2.
Violations Within the scope of this inspection, a cited licensee-identified slolation
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was noted, which concerned operation of the facility in cold shutdown for a period of about 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> without an operable containment radiation gaseous and particulate monitor.
(Section 8.3)
One'non-cited licensee-identified violation was identified for the event documented in LER 89-011-00.
(Section 8.3, page 25)
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3.
Unresolved Items Five previously unresolved items were closed, and two items were updated and remain open (Sections 5.3.1., 6.1.1., 6.1.2., 8.2.1., 8.2.2., 8.2.3.,
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and 8.2.4.).
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Four unresolved items were opened regarding: (1) pre planned alternate monitoring method for high range noble gaseous monitor (Section 4.3.2.);
(2) modification configuration control process as it relates to incorpor-ating vendor information (Section 8.2.4); (3) the consequence of failure to test control room annunciators (re: LER 89-008-00) (Section 8.3, page 20); (4) revision of LER 89-009-00 (Section 8.3, page 23); and,
(5) timeliness of reportability evaluations (re: LER 89-011-00) (Section 8.3, page 25).
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Allegation Followup A worker concern regarding the capability of the steam jet air ejector radiation mon'itor to detect a tube leak was reviewed.
(Section 4.3.1.)
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TABLE OF CONTENTS P.892 1.0 Persons Contacted..........................................
I 2.0 Summary of Facility Activities.............................
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3.0 Plant Operations (IP 71707/93702)*.........................
3.1 Control Room Observations.............................
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3.2 Plant Tours...........................................
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3.3 Review of Plant Incident Reports......................
4.0 Radiological Controls (IP 71707)...........................
4.1 Posting and Controls of Radiological Areas............
4,2 Chemistry Controls (Primary / Secondary)................
4.3 Radiation Monitor Alarm Issues........................
4.3.1 Steam Jet Air Ejector Radiation Monitor Performance...........................................
4.3.2 Operability Evaluation for High Range Noble Gas Effluent Monitor............................
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5.0 Maintenance / Surveillance (IP 62703/61726/92701)............
5.1 Observation of Maintenance Activities.................
5.2 Observation of Surveillance Activities.............
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5.3 Previously Identi fied Items...........................
5.3.1 (Closed) Unresolved Item 336/88-24-04:
Heise Pressure Gauge (0-5000 psig)
Acceptance Criteria...................................
5.4 Predictive Maintenance Pilot Program..................
6.0 Engineering / Technical Suppo: t (IP 92701/71707).............
6.1 Previously Identified Items...........................
6.1.1 (Closed) Unresolved Item 336/85-35-01:
"Li mi torque Ope ra tor Wi ri ng"..........................
6.1.2 (0 pen) Unresolved Item 336/89-18-01:
Implementation of NRC Guidance for the Anticipated Transient Without Scram Equipment Operability...........................
6.2 Reportability/ Justification for Continued Operation Evaluations.................................
6.2.1 Spare Battery Charger Output Breaker Operability Evaluation........................
7.0 Security (IP 717107/TI 2515/104)...........................
7.1 Inspection Tour Observations..........................
7.2 Fitness for Duty Program Initial Training Review......
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8.0 Safety Assessment / Quality Verification (IP40500/92702/92701).....................................
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8.1 Committee Activities..................................
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8.2 Previously Identified Items...........................
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8.2.1 (Closed) Violation 336/87-15-01:
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"Vulkene SIS Wire Deficiencies".......................
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8.2.2 (Closed) Unresolved Item 336/89-01-01:
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No Mechanism for Coordinating a Quality Assurance
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Overview of Safety-Related Activities.................
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8.2.3 (0 pen) Unresolved Item 336/89-18-02:
Untimely Response to Licensee's Quality Assurance Audits..............................
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8.2.4 (Closed)UnresolvedItem 336/89-13-05:
Verification of Reactor Coolant Pump Speed Sensing Cable Bend Radius.............................
8.3 Licensee Event Report Rev1ew..........................
8.4 Pe ri od i c Re po rt s......................................
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9.0 Management Meetings (IP 30703).........................-...
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- The NRC inspection manual inspection procedure (IP) or temporary instruction
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(TI) that was used as inspection guidance is listed for each applicable report
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DETAILS 1.0 persons Contacted Interviews and discussions were conducted with Northeast Nuclear Energy Company (NNECO or the licensee) staff and management during the report
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period to obtain information pertinent to the areas inspected.
Inspection findings were discussed periodically with the supervisory and management personnel identified below.
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- S. Scar.e, Millstone Station Superintendent
- J. Keenan, Unit 2 Superintendent
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J. Riley, Unit 2 Maintenance Supervisor J. Becker, Unit 2 Instrument and Controls Supervisor
J. Smith, Unit 2 Operations Supervisor
- Attendee at post-inspection exit meeting on January 26, 1990.
2.0 Summary of Facility Activities Millstone Nuclear Power Station (Millstone 2 or the plant) operated at l
rated thermal power throughout the inspection period.
On December 12, at approximately 8:00 p.m., the licensee identified an electrical ground on 125 vde vital inverter no. 1.
Upon further investigation, the licensee identified a service water leak on the header to the "B" emergency diesel generator.
The cause of the ground was service water leakage onto a motor-operated ventilation damper (2-EB-56).
Damper 2-EB-56 is an active component for engineered safety feature (ESF) systems (enclosure building filtration actuation signal, and auxfliary exhaust actuation signal).
The licensee's immediate corrective action:; included removal of power to the motor-operated ventilation damper, and isolation of the series damper to maintain the condenser air removal system isolated from the enclosure building filtration system and maintain the ventilation system in its required ESF alignment.
On December 13 at 10:00 a.m., the licensee isolated the service water to the "B" emergency diesel generator and declared it inoperable.
The leak area was ultrasonically tested, and repair activities were completed. At approximately 5:55 p.m. on December 13, the licensee declared the "B" emergency diesel generator operable, and on December 15, corrective maintenance activities were completed successfully on damper 2-EB-56.
On December 20, the utility identified a second vital 125 vde ground at pressurizer vent valve 2-RC-424.
Electric power to the valve solenoid was isolated and the technical specification requirement for pressurizer vent operability was met. The limiting condition for operation was maintained throughout the inspection period, for this component, with no inadequacies note *
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NRC Activities Routine resident inspection involving 131 regular hours, 9 back. shift hours, and 2 deep backshift hours.
3.0 Plant Operations 3.1 Control Room Observations Control room instruments were observed for correlation between channels, proper functioning, and conformance with technical speci-fications. Alarm conditions in effect and alarms received in the control room were discussed with operators.
The inspector periodi-cally reviewed the night order log, tagout log, plant incident report (PIR) log, key log, and bypass jumper log. The following tagouts were reviewed for acceptability and implementation: 2-3-90 "F-20 Heat Exchanger," 2-2-90 "C Instrument Air compressor," 2-4241-89
" Pressurizer Vent Valves 2-RC-424, 2-RC-425," 2-3022-89 " Spare Battery Charger," and 2-4193-89 " Charging system valve 2-CH-507." No inadequacies were noted.
Each of the respective logs was discussed with operation department staff.
No inadequacies were noted.
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3.2 plant Tours The inspector observed plant operations during regular and backshift tours of the following areas:
Control Room Containment Vital Switchgear Room Diesel Generator Room Turbine Building Intake Structure Enclosure Building ESF Cubicles During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and to verify correct communication and equipment status. No inadequacies were noted.
3.3 Review of Plant Incident Reports The plant incident reports (PIRs) listed below were reviewed during the inspection period to (i) determine the significance of the events; (ii) review the licensee's evaluation of the events; (iii)
verify the licensee's response and corrective actions were proper; and, (iv) verify that the licensee reported the events in accordance with applicable requirements, if required.
The PIRs reviewed were:
89-132, 'B' service water header leak 89-136, reactor protection system channel
'D' power supply failure
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,89-135, auxiliary exhaust actuation system radiation monitor RIT - 8157 out of calibration 90-03, RM-5099 isotopic out of calibration No inadequacies were noted.
4.0 Radiological Controls 4.1 Posting and Controls of Radiological Areas During plant tours, posting of contaminated, high airborne radiation, and high radiation areas were reviewed with respect to boundary identification, locking requirements, and appropriate control points.
No inadequacies were noted.
4.2 Chemistry Controls (Primary / Secondary)
The inspector reviewed primary and secondary chemistry results during the inspection period.
The review included a comparison of measured values to technical specification (TS) 3.4.7 " Reactor Coolant System Chemistry," TS 3.4.8 " Reactor Coolant System Specific Activity," TS 3.7.1.4 " Specific Activity Secondary Coolant," and Procedure OP 2217, Rev. 4, Secondary Chemistry Control, requirements.
No inadequacies were noted.
4.3 Radiation Monitor Status Alarm Issues 4.3.1 Steam Jet Air Ejector Radiation Monitor Performance The inspector reviewed the performance of the steam jet air ejector (SJAE) radiation monitor, and assessed the safety significance to determine whether licensee actions sufficiently addressed the operation of the radiation monitor.
Background As documented in inspection report 50-336/89-13 dated November 9, 1989, an allegation questioned the capability of the SJAE radiation monitor to detect a steam generator tube leak. The previous NRC review substantiated the allegation. The primary focus of this review is to assess further the need for and reliability of the SJAE radiation monitor.
The inspector reviewed the condition of the steam generator tubes from eddy current testing results during the February, 1989 refueling outage, and the mid-cycle outage conducted in November, 1989.
The review focused on the number of tubes identified with circumferential1y oriented cracks.
In February, 1989, 309 tubes with circumferential cracks were detected, and during the November, 1989 eddy current inspection an additional 104 cracks were identified.
In January,1987 the unit
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experienced a tube leak from a circumferential crack at the top of the tube sheet which forced a shut down of the plant. The event description was previously documented in inspection report 50-336/87-01. As documented in inspection report 50-336/89-13 the SJAE monitor did not indicate an increase in steam generator activity until two days af ter chemistry sampling, and af ter the steam generator blowdown monitor indicated a tube leak.
Final Safety Analysis Report section 14.14.1 states that
" detection of the steam generator tube rupture incident is facilitated by radiation monitors in the steam generator blow-down lines and in the condenser air ejector discharge lines.
The monitors initiate alarms in the control room and alert the operator of abnormal activity levels and that corrective actions are required."
performance During mid-December, the SJAE radiation monitors developed spurious spikes in radiation indication resulting in multiple isolations of the steam generator blowdown system. The inspector identified no inadequate control room operator actions as a result of the isolation and procedural follow-up actions.
Various plant operations review committee (PORC) meetings during the inspection period discussed the performance of the radiation monitor, verification of the activity in the SJAE samples, and initiation of corrective maintenance activities.
The inspector reviewed the previous corrective maintenance activities on the SJAE during power operations between 1988 and 1989. Approximately twenty-seven authorized work orders were performed by the licensee.
Effectively 14 days out of 329 days the radiation monitor was out of service solely for corrective maintenance, excluding required technical specification surveillances (i.e. source check, channel calibration, and channel function test).
The general deficiencies with the monitor performance were flooding, low flow alarms, no audible horn alarm, monitor spiking after trip monitor floods.
Past utility corrective actions included heat tracing the inlet lines to the radiation monitor to prevent flooding, and adjusting sample flow.
Recently (late December 1989/early January 1990), the SJAE monitor was out of service for troubleshooting based on numerous steam generator biowdown isolations.
In review of completed authorized work orders, three primary items in SJAE performance were identified by the licensee:
a potential dilution from the instrument air flow based on the isotopic gas calibration results; sample flow low (i.e. 1.8 standard cubic feet per minute (SCFM) vs. 3.0 SCFM; and inappropriate wire configuration for the detector establishing a potential ' floating' ground.
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The instrument loop folder indicates the flow rate and detector electrical installation were previously identified respectively
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in January and December, 1987.
i Chemistry Sampling As prescribed in procedure SP-2802 the required sampling frequency for SJAE exhaust is monthly for tritium and gaseous activity.
The licensee administratively completes the analyses t
weekly and in the event of a radiation level exceeding the alarm setpoint for the monitor.
Further guidance prescribed by the assistant chemistry supervisor per document MP-S-C-89-215 indicates a daily sampling frequency, consisting of: leakrate determination, trending of data, and establishment of a baseline.
The specific guidance requires a sample frequency of twice per day if the primary-to-secondary leakrate calculation
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increases to 5 gallons per day.
In the event the leakrate calculation exceeds 72 gallons per day, the sampling frequency is three times per day.
The technical specification 3.4.6.2.c
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limit is I gallon per minute (gpm) total primary to secondary and 0.10 gpm through any one steam generator. The leakrate determinations for the SJAE are based on the isotopes Xenon-133, Xenon-135, Xenon-135M for gaseous and iodine-131, iodine-133,
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and iodine-135 for particulate.
Safety _ Significance The safety significance of SJAE radiation monitor operation was assessed based on the emergency operating procedures, technical specification requirements and basis, Final Safety Analysis Report (FSAR), and integrated plant detection capabilities of a potential steam generator tube leak / rupture.
The emergency operating procedures (EOPs) use the SJAE radiation monitor alarm in the break identification chart to assess a primary coolant or main steam line rupture, and the monitor alarm is an entry condition for E0P 2534 " Steam Generator Tube Rupture." The SJAE radiation alarm, however, is not relied upon solely, as a decision point for control room operators actions.
Technical Specification 3.3.3.9 table 3.3-12 item Ic describes the requirements for operability of the SJAE radiation monitor or the steam generator blowdown radiation monitor.
Specifically, if both radiation monitors are inoperable, best efforts to repair the monitors are required, and in the interim chemistry grab samples of steam generator activity for gross radioactivity are acquired. The frequency of grab samples is based on reactor coolant system gross equivalent iodine.
No licensee action is required if one of the two monitors is inoperable.
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The FSAR description on the steam generator tube rupture event (Section 14.14) indicates that detection and integrated plant behavior varies depending upon the size of the tube rupture.
For leakage rates up to the capacity of the chemical and volume control system charging pumps, reactor coolant inventory can be maintained and an automatic reactor trip would not occur.
The gaseous fissien products would be released to the atmosphere from the main steam system via the condenser air ejector discharge to the unit I stack.
Those fission products not discharged in this way would be retained by the main steam, feedwater and condensate systems.
For leaks that exceed the capacity of the charging pumps, pressurizer water level and pressure decrease and an automatic reactor trip results.
The licensee's other methods of steam generator tube leak detection consist of, but are not limited to, the following parameters: steam generator blowdown radiation monitor alarm; main steam line radiation monitor, unbalance in the control of letdown / charging flowrates, pressurizer level and pressure decrease, start-up of stand-by charging pumps, and chemistry samples for gross radioactivity.
Conclusion Licensee efforts to address problems with the SJAE radiation monitor were extensive; however, the significant amount of corrective maintenance was not successful in assuring reliability of the monitor.
The inspector noted that detection of a steam generator tube leak is not solely related to the performance of the SJAE monitor; however, poor performance of this monitor removes a valuable source of information to the control room operators.
4.3.2 Oyerability Evaluation for High Range Noble Gas Effluent Monitor On January 2, 1990 at approximately 10:00 a.m., the licensee was initiating a required calibration of the high range noble gas effluent monitor (RM-8168).
The calibration is required by technical specification surveillance table 4.4-3 item Id.
RM-8168 failed its calibration acceptance criteria; specifically, the channel 3 particulate iodine Geiger-Mueller detector failed its 100 millirem / hour (110%) acceptance criteria. On January 12, the plant operations review committee approved an operability evaluation for RM-8168 based on the unsuccessful calibration.
Description RM-8168 (Kaman channel) is the in-line accident range stack gas i
effluent radiation monitor.
It samples the Unit 2 stack by one
of three isokinetic nozzles. The monitor has five channels for
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operation.
Channels one and two measure gaseous activity, and channels three, four, and five are particulate iodine collectors. Channel one is the mid-range gas channel with a range between.001 to 50 microcuries / cubic centimeter (pCi/cc), and the high range gas channel covers between 5.0 to 10,000 pCi/ce.
The high range alarm is activated by channel 1 at.2 pC1/cc.
The description of RM-8168 is in Final Safety
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Analysis Report (FSAR) 7.5.6.3.2.1.1 " Unit 2-Stack Gaseous and Particulate Monitoring."
NUREG-0737, Clarification of TMI Action Plan Requirements, describes the design basis of the monitor. The basis includes prompt quantification (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) of certain radionuclides that are indicators of core damage (i.e. noble gases [fueldamage),oriodines[highfueltemperatures]).
The basis to obtain and analyze the sample, is to not exceed radiation exposures to an individual, as required in general design criterion 19 of 10 CFR 50, Appendix A.
NRC Regulatory Guide 1.97 " Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Condition During and Following an Accident" prescribes the effluent noble gas monitor as a Type C, Category 2 instrument.
The purpose of the monitor is to provide an indication of a breech or actual breech of the barriers to fission product release. Category 2 does not generally include seismic qualification, redundancy, or continuous display, but requires a high-reliability power source.
At Millstone 2, the emergency classifications in part prescribed in emergency plan implementing procedure (EPIP) 4701-2 are based on RM-8168 gaseous indications.
The following is the classification and RM-8168 indication: Alert (.03pC1/cc -
.3pci/cc); Site Area Emergency (.3uct/cc - 1.0 uti/cc) and Gensral Emergency (1-6 uCi/cc).
EPIP 4217 " Vent and Containment Air Post Accident Sampling" section 3.3.2 describes the methods to determine and assess the filter / charcoal canisters and provide the results to the Manager of Radiological Consequences Assessment in the event of an emergency.
EPIP 4217 step 3.3.2.43 prescribes utility actions if the filter activity (channels 3, 4, 5 of RM 8168) are in excess of 100 millirem / hour.
Requirements of RM-8168 Technical specification requirements for this radiation monitor are identified in 3.3.3.1, Table 3.3-6 item Id.
The requirements are that the monitor be continuously operable during plant operational modes 1-4, with an alarm trip setpoint of 0.2 pCi/cc for the gaseous channe w..
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The required actions if the monitor is inoperable are to initiate preplanned alternata methods within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and restore the channel to operable status within 7 days of discovery or prepare and submit to the NRC a special report pursuant to requirement 6.9.2 within 14 days of discovery.
RM-8168 was taken out of service by the licensee on January 2 and declared operable based on the operability evaluation on January 12. At the end of the inspection period, the licensee was preparing a special report pursuant to TS 6.9.2.
The particulate sampling capability is required by TS 6.18
" Post-Accident Sampling System".
No specific equipment requirements are specified.
Procedures for sampling are SP 2863 and CP 2810N, and EPIP 4217.
The filter / charcoal composed of silver-zeolite material are replaced monthly with no analysis required. The replacement is based on shelf-life of the silver-zeolite material.
Pre-Planned Alternate Methods Inspector review of licensee actions to meet the technical specification on January 8 identified three deficiencies.
The alternate pre planned method for monitoring the stack high range effluents consist of a portable area monitor (Dosimeter Corporation Model 3090-3 with Model 30947-R-100 type probe)
located at the 36-foot elevation in the east penetration room.
Once a shift, a plant equipment operator reported the monitor readout to the control room. Operators recorded the readings in a log.
The channel readout was provided on a logarithmic scale on the face of the monitor in units of Roentgen per hour (R/hr) over a range from 1 mR/hr to 100 R/hr. The detector probe was strapped to the underside of the ventilation outlet duct using plastic tie wraps and the monitor was likewise strapped to unistrut on a wall about 15 feet from the probe. The readings on January 8 were about 30 mR/hr. A cesium-137 source strapped to the probe contributed 20 mR/hr to the normal reading. The monitor had a calibration sticker affixed indicating that it was within the specified calibration interval.
When asked by the inspector to demonstrate use of the alternate monitor in conjunction with the EPIPs, the operator was not able to make a classification using the alternate method since the EPIPs were written in units of uC1/cc and no correlation data were available to allow conversion from R/hr to microcuries.
Inspector inquiries with personnel in the operations, site engineering, health physics and chemistry departments determined that the needed correlation data was not contained in the EPIPs or otherwise available onsite.
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When the matter was discussed with site engineering personnel, the licensee stated that the lack of correlation data for the channel was not significant, since if plant conditions were degraded to the extent that the monitor was needed, the emergency response facilities would be staffed and the classification function would be provided by emergency dose assessment personnel. While the inspector agreed that the classification function could be provided by the emergency
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organization once the emergency response facilities were fully
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staffed, the " alternate monitoring method" in place on January 8 could not be used in a meaningful way by site personnel and
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could not be used in place of the Kaman monitor as intended to
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provide "immediate" classification capabilities using control
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room indications.
I In response to inspector concerns, a copy of a correlation curve was obtained from the corporate engineering office on January 8 and made available to the control room.
The stack low range monitor (RM 8132) provides measurement of stack releases over a range of IX10-6 to IX10+4 uti/cc.
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Kaman channel provides overlap with the low range and measures releases over a range of IX10-3 to IX10+5 pCi/ce. The calibration curve provided for the alternate monitor showed a measurement range of 0.15 uti/cc (=0.1R/hr) to 1.5X10+2 pC1/cc (100 R/hr). Thus, at the top end of its range, the alternate
"high range" stack monitor did not provide coverage equivalent to the Kaman channel and essentially replicates the range of the original " low range" monitor.
The monitor range, however, was adequate to essentially cover the emergency classification cut-of f levels of from 0.03 pCi to 3 pCi/cc.
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To obtain data from the alternate monitor as presently configured, an operator would have to enter the east penetration room to obtain local indication from the face of the meter.
This conceivably put an individual within 10 feet from a 100 R/hr source term, assuming the monitor was reading at full scale. Even though excess lengths of cable (estimated by the inspector to be dozens of feet) was provided between the monitor and the probe, it was wrapped in a coil near the monitor. The extra cable could be used to increase the distance of the monitor from the probe and thus reduce the dose to personnel.
Effective use of the extra cable, however, would require constructing a penetration through a double door air lock between the penetration room and the spent fuel pool area.
Inspector review noted that Millstone 2 TS 3.3-6 conforms with the standard technical specification requested by the NRC staff for stack noble gas high range monitor as part of the TMI action plan. Neither the standard specification nor the Millstone 2
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bases provide guidance on what requirements should be placed on the " alternate monitoring method".
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The inspector concluded this matter requires further action by the licensee to assure plant personnel can comply with the TS
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3.3-6 action statement at times when the normal high range I
(Kaman) monitoring channel is unavailable. This item is unresolved pending action by the licensee to improve procedures
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and controls for the alternate monitoring method, and subsequent i
review by the NRC (336/89-24-01).
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f Review of Operability Evaluation i
The inspector reviewed the licensee operability evaluation
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consisting of:
the identified deficiency, affected system,
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discussion / consequence, and conclusions. The inspector concluded that the calibration failure for channel 3 i
(particulate / iodine cartridge) did not impact control room i
operators' ability to classify emergency events (radiological consequences) or develop a potential to exceed criteria 19 of 10 CFR 50 Appendix A to obtain/ analyze the cartridge, or affect the module alarm. The emergency events are classified using channel
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1 or 2 (gaseous), and the filter trip function was calibrated less than the prescribed 100 millirem / hour radiation field.
The operability, as prescribed in the applicable technical specification basis for RM-8168 channels, ensures that the
radiation levels are continually measured in areas served by the
individual channels and the alarm / automatic action is initiated
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when the radiation level trip setpoint is exceeded.
Based on the calibration results, the channels continually measure gaseous and particulate levels, and the alarm setpoint (0.2 pCi/cc) was acceptable (by calibration); however, the setpoint
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(100 millirem /hr) to switch cartridge collectors sequentially was out-of-specification low (i.e. 70 mr/hr) for channel 3.
FSAR section 7.5.6.3.2.1.1 describes the purpose of the 100 mr/hr setpoint is the maximum amount of radioactivity to be handled by ene person safely.
The inspector had no further questions on the operability evaluation and in future inspections will verify the licensee actions to repair RM-8168.
5.0 Maintenance / Surveillance 5.1 Observation of Maintenance Activities The inspector observed and/or reviewed selected portions of preven-tive and corrective maintenance to verify compliance with regulations, use of administrative and maintenance procedures, compliance with codes and standards, proper QA/QC involvement, use of
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bypass jumpers and safety tags, personnel protection, and eq,ipment
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alignment and retest. The following activities were included:
AWO M2-89-14679 Channel D Reactor Protection System 18 VDC Power
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Supply Replacement on January 2
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AWO M2-89-09860, Annual Preventive Maintenance on "C" Instrument
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Air Compressor j
No inadequacies were identified.
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I 5.2 Observation of Surveillance Activities I
The inspector observed portions anci/or review of completed surveil-
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lance tests to assess performance in accordance with approved pro-
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cedures and Limiting Conditions of Operation, removal and restoration l
of equipment, and deficiency review and resolution. The following
tests were reviewed:
SP 21108 Rev. 5, Reactor Building Component Cooling Water
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(RBCCW) Pump ' A' Operational Readiness Test i
SP 21109 Rev. 5, Reactor Building Component Cooling Water
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(RBCCW) Pump 'B' Operational Readiness Test SP 2604E-3, Facility I "High Pressure Safety Injection
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Operability Test" In-Service Test 89-25 Loss of i 18 VDC power to Core Protection
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Calculator #2 for the 'D' Reactor Protection System Channel No inadequacies were noted.
5.3 Previously Identified Items 5.3.1 (Closed) Unresolved Item 336/88-24-04: Heise Pressure Gauge (0-5600 psig) Acceptance Criteria As documented in report 50-336/88-24 dated November 23, 1988 the inspector noted a discrepancy in the licensee's application of the acceptance criteria of procedure I/C 1104A "I&C Pressure Test Gauges Calibrations." Paragraph 2 of I/C 1104A defines acceptance criteria for Heise gauges as 10.1% of full scale or i the minor scale division, whichever is greater.
Heise gauge No.
QA-370 is a 0-5000 psig (Model CMM) gauge with 5 psi minor scale divisions.
The inspector noted that the acceptance criterion used for calibration of this gauge was 110 psi, which is inconsistent with the required acceptance criteria of 15 psi.
On January 25, 1989 the licensee's site operations review committee (SORC) approved a procedural form change to I/C 1104A.
Form 1104A-41 was added to the procedure to include an acceptance criterion of 15 psi for Heise CMM six-inch pressure gauges (i.e. No. QA-370).
The inspector verified implementation
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of the procedural change form to I/C 1104A. This item is closed.
5.4 predictive Maintenance Pilot Program The inspector reviewed internal document MM-90-002 dated January 4, 1990 which describes the maintenance department's implementation of a predictive maintenance pilot program.
The purpose of the program is to monitor degradation of selected rotating equipment through vibrational analysis.
Fifty-seven components were selected for the program consisting of safety-related, important-to-safety, and balance-of plant equipment.
Selected examples of rotating equipment included in the program are:
instrument air compressors, control room air conditioning fans and compressors, reactor building component cooling water pumps, control element drive mechanism motor generator sets, emergency diesel generators, feedwater pumps, fire pump, and refuel water purification pump.
The vibration analysis uses a Technology for Energy Company (TEC) data collection meter and Intelli-Trend vibration analysis software installed on a personnel computer station.
Vibration blocks were installed on the horizontal, vertical, and axial direction of
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the associated equipment rotational centerpoint to achieve a reproducible data point.
Each selected component has a target vibration level established, of which the components are categorized in four ratings:
very low vibration, normal vibration, slightly rough, and very rough.
The very rough machines are monitored weekly.
Corrective, preventive maintenance, or design changes are implemented to bring the component within a normal vibration level.
The program is independent of the required in-service inspection (ISI); however, some equipment selected is under the ISI program.
Licensee corporate reliability engineering will review the collected data on a monthly and annual basis to determine data trends with recommended repair recommendations to the Millstone 2 maintenance department.
The program will be controlled and documented in the automated work order system.
The inspector considered that the establishment of a predictive maintenance pilot program for rotational equipment is a noteworthy initiative.
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6.0 Engineering / Technical Support 6.1 Previously Identified Items 6.1.1 (Closed) Unresolved Item 336/85-35-01: Operator Wiring" This item involved the licensee's failure to demonstrate qualification of internal wiring in Limitorque valve operators.
Previous NRC inspection 50-336/87-15 close-out of this issue was based on licensee valve walkdowns to verify equipment qualification including terminal block concerns addressed in NRC Information Notice 83-72, and NRC inspection of the walkdown data. Hence, this item is being closed administrative 1y in this report.
6.1.2 (0 pen) Unresolved Item 50-336/89-18-01:
Implementation of NRC Guidance for the Anticipated Transient Without Scram (ATWS)
Equipment Operability The status of this issue was reviewed with the licensee.
This item remains open pending NRC issuance of guidance to the licensee and the licensee taking action regarding the implementation of the ATWS equipment operability and surveillance requirements.
6.2 Reportability/ Justification for Continued Operation Evaluations 6.2.1 Spare Battery Charger Output Breaker Operability Evaluation As documented in NRC inspection report 50-336/89-22 dated January 3, 1990 the inspector requested plant management to provide a justification for continued operation for a refurbished overload trip unit based on information contained in NRC Information Notice 89-45.
The initial engineering' evaluation at Millstone 2 identified one DC overload trip unit refurbished by Satin America (ref.
Information Notice 89-45) that is currently installed in a IE breaker.
The breaker is for the " spare" battery charger and is not normally in service.
The breaker is an AK-2A-50-1 type with an EC-1 trip device. The EC-1 trip device is a protective relay that acts to initiate fault current interruption as a function of time and current values.
The EC-1 trip device was refurbished by the Satin America Company in 1985.
NNECO prepared and approved the operability evaluation on November 8, 1989, as documented in plant operations review committee meeting 2-89-178. The evaluation was comprised of L
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seven parts, including:
initial condition; system design basis; i
system operability with identified deficiency; corrective
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action; operator awareness; and duration. As documented in the evaluation, the EC-1 devices provide protection to the swing battery bus D2010, which is normally de-energized and only in service as a result of failures to the "A" or "B" battery I
t charger.
The inspector review indicated that the EC-1 trip device
acceptance criteria were met for initial plant startup and after refurbishment from Satin America (AWO M2-850797).
In June,1989 the licensee, in response to NRC Information Notice 89-45, performed the instantaneous trip current test of procedure PT-1452B. The results of the surveillance failed the instantaneous pick-up acceptance criterion of 110%.
Specifically, the instantaneous trip specification is nine times the breaker coil rating or 7200 amps.
However, the result recorded for the negative pole was 5500 amps. The time to trip (acceptance criterion less than or equal to 0.05 seconds) at 150%
of instantaneous current (10,800 amps), the long time, and short time delay test results were acceptable.
In response to the discrepancy, the licensee initiated non-conformance report 2-89-150. The engineering staff determined that the instantaneous trip specification outside the acceptance criterion was acceptable, based on the battery charger o
maximum current limit of 1,000 amperes.
The only condition for the output breaker to exceed 1,000 amperes is a charger fault.
The instantaneous current setting for the battery charger output breaker is not considered as part of the vital 125 VDC breaker coordination scheme.
At the end of the inspection period, the licensee was making preparations to replace the EC-1 trip devices with General Electric devices. The inspector reviewed the operability evaluation, previous corrective maintenance activities, and the most recent protective relay test and concluded that continued operation with the installed EC-1 devices is acceptable until completion of the replacement of General Electric supplied devices. The inspector had no further questions in regard to the licensee disposition and the operability evaluation.
7.0 Security i
7.1 Inspection Tour Observations Selected aspects of site security, including site access controls, personnel searches, personnel monitoring, placement of physical l
barriers, compensatory measures, guard force staffing, and response
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to alarms and degraded conditions, were verified to be proper during inspection tours.
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7.2 Fitness for Duty Program Initial Training Review l
The licensee implemented a revised fitness for duty (FFD) program to meet 10 CFR 26 which required the new rule be implemented before l
January 3, 1990. By letter dated December 18, 1989, the licensee certified the new program was implemented and meets the requirement J
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of 10 CFR 26 as currently understood. The December 18 submittal described the major facets of the new program, along with the drugs tested for in the program and the associated cut-off limits.
The
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licensee program tests for a more extensive list than is required by.
L the rule, and, in the case of marijuana, has a more stringent cutoff level.
The purpose of this inspection was to review the initial licensee
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training programs for the new FFD rule. Tne inspection was completed on December 21, 1989 in accordance with Temporary Instruction (TI)
L 2515/104 for two of the three parts of initial training: policy
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awareness training for employees and contractors, that will be
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included in the licensee general employee training program; and, training for escorts. The licensee also conducted training for
supervisors, which included general awareness training of the requirements of the rule, and also incorporated a two-day workshop to demonstrate behavioral observation techniques to detect substance abuse.
The supervisor training program was completed several months
prior to implementation of the revised FFD program and was not reviewed during this. inspection period. A make-up session, scheduled for licensee supervisors during the first quarter in 1990, is subject to inspection at that time.
The results of the training inspection were summarized on checklists provided in TI 2515/104, and were provided to the Office of Nuclear Reactor Regulation (NRR) for its review. The quality of the training programs was very good.
NRC review of the new FFD program for conformance with 10 CFR 26 will be completed on a subsequent routine inspection.
8.0 Safety Assessment / Quality Verification
8.1 Committee Activities The inspector attended plant operations review committee (PORC)
meetings 2-89-200, 2-89-202, 2-89-204, 2-90-01, 2-90-03, 2-90-04, and 2-90-06, on December 11, December 13, and December 19, 1989, and on January 2, January 4, January 5, and January 12, 1990. The inspector noted that committee administrative requirements were met for the meetings, and that committee functions were discharged in accordance with regulatory requirements. The inspector observed adequate discussion of matters before the PORC and a good regard for safety in
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the issues under consideration by the committee.
Selected topics for l.
discussions that were monitored, included:
i revision to LER 89-004 on mechanical steam generator plug
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repair.
technical specification procedure review as prescribed in
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corrective actions for LER 89-008, l
steam jet air ejector performance,
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emergency operating procedures (to Revision 3 of Guidelines
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CEN-152),
"A" emergency diesel generator corrective actions during the
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mid-cycle outage, and high range gaseous stack radiation monitor operability.
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No inadequacies viere identified.
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8.2 Previously Identified Items 8.2.1 (Closed) Violation 336/87-15-01: "Vulkene SIS Wire Deficiencies" l
As documented in inspection report 50-336/87-15, the inspector identified a violation of 10 CFR 50.49(f), in that qualification of each component was not based on testing or experience with identical equipment or with similar equipment with a supporting analysis to show that the equipment to be qualified was acceptable.
Specifically, thirteen motor-operated valves utilized Vulkene SIS wire and licensee-referenced qualification reports (Franklin test report F-C4497-2) applied to SIS Vulkene
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Supreme wire.
The utility responded to the notice of violation by letter dated December 18, 1987.
The licensee documented that all Vulkene SIS wire was replaced wit.h environmentally qualified wire during the July, 1987 inspection.
The inspector verified that the replacement of unqualified wire was accomplished by review and discussions of authorized work orders completed in July,1987.
The specific valves affected were 2-CS-16-1A, 2-SI-656, 2-SI-646, 2-CH-501, 2-SI-654, 2-SI-635, 2-RB-37-2A, 2-RB-37-2B, 2-RB-30-1A, 2-RB-30-1B, 2-51-651, 2-51-614, and 2-SI-644 located in the safety injection, charging, containment spray, and reactor building component cooling water systems.
The licensee was able to establish similarity of wire (i.e. verified Vulkene Supreme SIS) in valves 2-CS-16-1A, 2-51-656, and 2-51-646 in the July,1987 walkdown, thereby taking credit for test report F-C4497-2).
Wyle test report No. 47839-02 is a qualification report for Vulkene SIS wire tested in a penetration. On August 11, 1987 the licensee prepared GJP-87-295 engineering evaluation to compare the Wyle test report to determine the qualification level of the Vulkene SIS cable to the environmental parameters I
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set forth in the NNECO EEQ file MP2-122.
The evaluation conclusion determined that the various environmental tests
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demonstrated the ability of the Vulkene SIS cable in Limitorque
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motor-operated valves to remain within.the parameter guidelines
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described in the NRR Guideline for Evaluating Environmental
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Qualification of Class IE Electrical Equipment in Operating Reactors. The inspector reviewed the engineering evaluation with the Wyle test record and developed no further questions.
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In conclusion, the licensee replaced the unqualified wire in r
nine motor-operated valves, evaluated the initial configuration, and determined the Vulkene SIS wire was qualifiable from the
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10CFR 50.49 requirements.
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The licensee actions to prevent recurrence of a difference in qualification test reports and as-built applications included:
counseling engineers to closely scrutinize possible model/ type differences between installed equipment and qualification files,
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participation in industry EQ group efforts, and a review of the EQ program to assure the discrepancy was an isolated issue.
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This item is closed.
8.2.2 { Closed)UnresolvedItem 336/89-01-01: No Mechanism for Coordinating a Quality Assurance Overview of Safety-Related Activities
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The station quality assessment services supervisor presented to the inspector his 1990 integrated assessment plan as evidence
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that the licensee is integrating and coordinating the quality services department (QSD) verification activities.
The coordination effort was supported by the plant quality services l
manager and supervisor. The plan described the methods for performing the assessment function, i.e., audits, surveillances,
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and inspections; the major activities to be assessed; and i
included a description of how the assessment activities will be accomplished during 1990.
The inspector reviewed the revised ACP-QA 9.070, " Quality t
l Services Surveillance Program" that incorporated the response requirements regarding surveillance findings. The automated
surveillance tracking log was also reviewed. Only one overdue item was listed for the station as compared to an August 16, 1989 report that indicated they had 15 overdue items.
l Based upon these findings, the item is closed.
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18-n 8.2.3- (0 pen) Unresolved Item 336/89-18-02: Untimely Response to Licensee's Quality Assurance Audits Licensee actions taken to correct the untimely response to licensee's QA audits included the station superintendent's memorandum of August 10, 1989 and the revision of the Nuclear Engineering and Operations Procedure (NEO) 3.07, " Response to Audit Findings "
NEO 3.07 became effective on February 20, 1990. This item remains open pending the verification of NEO 3.07 implementation, 8.2.4 (Closed) Unresolved Item 336/89-13-05: Verification of Reactor Coolant Pump Speed Sensing Cable Bend Radius TH. :.em was open pending NRC verification of as-found minimum beg;;i.g radii for the microdot coaxial cable as prescribed by the manufacturer (3 inches for the probe lead and 1.5 inch for the extension cable) for the reactor coolant pump (RCP) speed sensors.
On November 16, 1989 the inspector received two concerns from a Northeast Utilituts employee associated with: (1) the status of NRC verification of-the bend radius of microdot cable for RCPs and (2) use of aluminum electrical metallic tubing (EMT) for the application of the new speed sensor assemblies installed in the mid-cycle outage.
On November 17, the inspector reviewed the A, B, and D reactor coolant pump speed sensing cable bending radii for the probe lead and extension cable in the vicinity of'the RCP motor af ter modification under MP2-89-114 "RCP Speed Sensor Conduit Improvements". The C RCP was not verified due to-inaccessibility (no staging). The utility documented the as-found bending radius was exceeded for the probe lead on the C and D RCP in non-conformance report 290-001. Section 2-16 of vendor technical manual (VTM2-150-012A) provides a recommended minimum bend radius for the microdot coaxial cable. As documented in inspection report.50-336/89-13 the C and D RCP motors were replaced in the 1986 and 1989 refuel outages. The as-found configuration confirms for both installations the recommended vendor installation guidelines were not
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incorporated.
The above item is closed, however, the licensee's modification configuration control process, specifically as it relates to' incorporating guidance from vendor manuals, is unresolved, pending subsequent NRC resident inspection (336/89-24-02).
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As documented above, the inspector verified the bend radius for the speed cables of three RCPs after the modification and reviewed the licensee's non-conformance report on the as-found configuration.
PDCR evaluation MP2-89-114 documents the new conduit configuration and the engineering calculation MP2-89-114-1226-GP Rev. 00 for seismic acceptability. The conduit installation was found acceptable for deadweight, operational base earthquake, and safe shutdown earthquake loads, based on inspector review of Section 5.0 of the Final Safety Analysis Report (FSAR).
The EMT conduit was replaced with a rigid conduit.
In both
' cases the conduit material was not composed of aluminum.
FSAR Section 14.18.2.3 " Corrosion of Containment Metal" specifies the amount of aluminum inside containment based on the corrosion effects of containment spray and the liberation of free hydrogen.
This issue is closed.
8.3 -Licensee Event Reports Review Licensee event reports (LERs) submitted during the period were reviewed to access LER accuracy, the adequacy of corrective actions and compliance with 10CFR73 reporting requirements, and to determine if there were any generic implications or if any-further information was required.
Four LERs were reviewed:
LER 89-008-00:
" Incomplete Surveillance Requirements for Testing
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of Annunciators Circuits" On October 2, 1989 a NNEC0 management review of a reportable occur-rence submitted for another facility (Haddam Neck) determined that Millstone 2 was not performing a functional check of control board annunciators during channel functional tests, as was required by the facility technical specifications. NNEC0 documented that the root cause of this event was personnel error in initial generation and interpretation of technical specification surveillances.
Its operation was not ' verified as an acceptance criterion.
The licensee presently believes that the remote alarm circuitry serves no safety function, but has concluded the failure to test the alarm function constitutes a technical specification compliance issue. The inspector considers the remote alarm circuitry important to safety since it provides information to operators on the status of plant equipment, and is used in the emergency and operating procedures to direct operator actions to mitigate off-normal
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conditions, transients and accidents. The control room alarm:, serve a safety function in control and operation of Millstone 2.
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The licensee reported the LER 89-008 under 10CFR50.73(a)(2)(1)(B), as a condition prohibited by the plant technical specifications, on
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November 1, 1989. The Haddam Neck LER was documented in August, 1989.
During the months of September and October, 1989 NNECO had completed a review of approximately twenty out of a total of eighty-four surveillance procedures to determine which required revision to incorporate checks of the control room annunciators.
l On November 29, 1989 the instrument and controls department developed departmental instruction 1.07 " Surveillance Compliance Verification" to proceduralize a method to periodically verify the existing surveillance procedure accuracy and compliance to the intent of technical specifications.
The inspector reviewed the status of utility corrective actions as documented in LER 89-008-00 as of December 21, 1989. Of the 84 instrument and controls department required surveillance procedures; 63 required no corrections; 14 were still under review / evaluation; and 7 required procedural changes to test control room annunciators.
The inspector discussed the status of corrective actions for LER 89-008-00 with licensee personnel and identified. additional licensee actions necessary to address NRC concerns. To assess the consequences of failure to test control board alarms the inspector requested the licensee to provide a status of alarms not verified and if any alarm circuits were identified as functionally inoperable based on a past failure to surveil. The inspector also noted that the surveillance procedure review was assigned to a single contractor person, with no second level checks performed by licensee personnel to assure completeness and adequacy of the surveillance procedure review effort. These itt.as are considered unresolved and will be addressed in future NRC inspection (336/89-25-03).
The other unit departments (i.e. chemistry, operations) had been advised by licensee management of the definition of channel functional test and are reviewing procedures under their cognizance to assure that technical specification requirements are met.
NNEC0 has committed to completion of this review by February 28, 1990, and committed to provide an update LER.
LER 89-009-00:
Radiation Monitor RM 8262 Inlet Valve (2-AC-82)
Found Closed On November 22, 1989 the licensee documented LER 89-009-00 pursuant to 10CFR50.73(a)(2)(1). On October 25, at approximately 6:49 p.m.
the plant was in cold shutdown (Mode 5). The licensee identified the
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containment purge isolation system (CPIS) was in operation (i.e.
purge valves were open) without operable radiation monitors. This event was reported as a violation of technical specification (TS)
requirement 3.3.3.1 " Radiation Monitoring." The acticn statement for
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3.3.1.1 requires the use of a constant air monitor or a grab sample at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
There are two CPIS channels with each channel consisting of a gaseous and particulate radiation monitor.
The engineered safety feature logic of (1 out of 4) automatically close the containment purge valves. The licensee reported all radiation monitors were out of service for 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> prior to identification of the inoperability.
Specifically, radiation monitor channel 8123 was out-of-service on October 24 at 1:10 p.m. for required calibration in accordance with procedure SP-2404AL.
Channel 8123 was restored to service at approximately 6:00 p.m. on October 25.
The redundant channel (RM 8262) was inoperable during the calibration of RM-8123 based on
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inadequate restoration from procedure SP 2404AL commencing on October 21, with the sample inlet valve (2-AC-82) in the incorrect position
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(closed).
Licensee Action and LER Deficiencies The inspector's review of the LER indicated two deficiencies.
First, the licensee failed to identify a second requirement for the CPIS.
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The requirement is stated in TS 3.3.2.1 Table 3.3-3, Item 7A, Action 3 in the engineered safety features section. The required action in cold shutdown (mode 5) is to immediately close the four containment purge valves upon discovery of no operable radiation monitors.
Second, the LER specified a period of 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> out-of-service for both channels of CPIS, whereas inspector review indicated approxi-mately 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />.
The LER did not address the issues that (1) control room operator action was required to close the containment purge valves, and (2) technical specification limiting conditions for-operation 3.3.3.1 and 3.3.2.1 were both exceeded.
Safety Impact The technical specification basis for the containment airborne radioactivity monitors are to automatically initiate closure of the containment purge valves upon detection of high airborne radioacti-vity levels in containment.
Closure of the purge valves prevents excessive radioactivity from being released to the environs in the event of an accident.
Final Safety Analysis Report section 7.5.6.3.2.1.2 documents that high radiation signals from any one of four detectors will result in an engineered safety features actuation to initiate a containment purge isolation.
During the period of inoperable radiation monitors, the plant was in cold shutdown and in reduced inventory operations (ref. NRC Generic Letter 88-17). The level in the reactor coolant system varied between 49.5 inches to 5 inches above the centerline of the reactor coolant system hotleg.
The B low pressure safety injection pump was
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in service for shutdown cooling operations.
As documented in routine
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inspection report 50-336/88-28, licensee calculations in the event of loss of shutdown cooling with no injection predict, for worst case assumptions, the on-set of core boiling in 13 minutes and core uncovery_ in 122 minutes with a 72-hour decay time.
The actual decay time was 87 hours0.00101 days <br />0.0242 hours <br />1.438492e-4 weeks <br />3.31035e-5 months <br /> when both channels of CPIS were identified inoperable.
Reduced inventory precautions for containment integrity were established and in place.
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In conclusion, the automatic isolation feature for the purge valves on high radiation was inoperable for approximately 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />, while in reduced inventory shutdown plant operation with a relatively high l
decay heat rate.
Mitigation Factors l
The mitigation factors present during the event are as follows:
(1) Reduced inventory actions for containment closure were established; I
(2) As documented in LER 89-009-00, the health physics organization conducted routine containment airborne samples with no abnormal results; (3) The-local and remote alarms for the area radiation monitor on the refuel bridge was operable during the event; (4) No potential source term from refueling activities as none existed during the event time; (5)- In the event of a loss of shutdown cooling, abnormal operating l
procedure ( AOP) 2572 step 4.1 directs the control room operators to verify containment penetrations are isolated and, if at 190 degrees F in the reactor coolant system or upon 15 minutes loss
of shutdown cooling, to evacuate containment.
Corrective Actions The root cause of the event was technician error in processing the
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initial procedure change, followed by insufficient surveillance procedure change controls.
Specifically, change 1 to procedure SP 2404AL, " Containment Gaseous / Particulate Process Radiation Monitor Calibration" relocated the isolation of the monitor outside the skid; however, the procedure without the change isolated the monitor within the skid (valve 2-AC-82).
The original procedure and change were utilized during the course of the surveillance. Restoration of the calibration left valve 2-AC-82 in the closed position and the licensee documented the cause as personnel error, whereas the s
inspector identified the cause as implementation, review and approval
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of the procedure change.
Specifically, the Plant Operations Review
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Committee approved change 1 to procedure SP-2404AL on October 18 in meeting 2-89-160 during the calibration.
The corrective actions for this event included procedure Change #2 for SP-2404AL (RM-8123) and Change #1 to SP-2404AM (RM-8262) to include a restoration step to double verify valve line-ups inside/outside the monitor skid, immediate opening of 2-AC-82 upon discovery of the misposition, and discussions with the instrument and controls department on lessons learned in making procedural changes prior to restoration of the system. The inspector verified the corrective actions for implementation and had no further questions.
The inspector concluded that the event as described in LER 89-009-00 is a violation of technical specification 3.3.2.1, as well as 3.3.3.1, and had low safety significance (336/89-24-04).
The LER incompleteness documentation, potential plant vulnerability, and " root cause" deterM nation to prevent recurrence is unresolved pending revision of tne LER to address the above inadequacies in the description of the event (336/89-24-05).
LER 89-010-00:
" Inoperable Service Water Strainers:"
Previous NRC aview of this event is documented in inspection report 50-336/89-23. The inspector has no further questions in review of LER 89-010-00, except for the unresolved item documented in report 50-336/89-23 on the acceptability of the licensee's seismic anchor assemblies.
LER 89-011-00: '" Service Water Isolation Valve - Incorrect Air Supply Check Valve Location" Background-On September 6,1989 at approximately 2:45 p.m., the licensee identified an incorrect check valve orientation for the instrument air supply to valve 2-SW-3.2A, which is a solenoid and air-operated sixteen-inch butterfly valve.
Valve 2-SW-3.2A is the service water supply isolation valve to one of two turbine component cooling water heat exchangers.
The licensee reported this event pursuant to 10 CFR 73(a)(2)(1) due to an operation prohibited by technical specifications.
The LER was dated January 5,1990.
The valve is designed to fail as-is on a loss of instrument air,and to close on a safety injection actuation signal. Reportability evaluation 89-43 determined that a loss of coolant accident, concur-rent with a loss of normal power would result in a significant diversion of service water flow from safety-related components cooled
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by this service water header based on the "as-found" air-supply check valve configuration.
The incorrect air supply check valve configuration was identified by a NUSCo probabilistic risk assessment engineer during a walkdown of the service water system. The licensee documented that the cause of the misconfiguration occurred during re-assembly of the instrument air line to the solenoid valve and accumulator during the February, 1989 service water pipe replacement. The service water header was returned to service on March., 1989.
Corrective Action The check valve was relocated by 4:45 p.m on September 6, and during the corrective maintenance, valve 2-SW-3.2A was maintained in operator manual control.
To prevent reoccurrence, departmental training for work supervisors and engineers will be implemented to be made aware of the event. The utility has committed to an update LER by April 30, 1990 to reflect completion of training.
Assessment The inspector reviewed both the technical content and timeliness of reportability evaluation 89-43 as prescribed in procedure NE0 2.25.
The NNECo technical review of this event considered the design basis accident with required flowrates to safety-related components and the configuration of valve 2-SW-3.2A, the loss-of-normal power coincident with a seismic event, service water pump flow, and required service water flow based on emergency diesel generator performance. The technical input into the reportability evaluation was comprehensive with good consideration to plant impact and safety.
With regards to the timeliness of the reportability evaluation, specific time intervals are: the event was identified and corrected on September 6, 1989; the reportability evaluation commenced on September 27 and concluded on November 21; the event was determined reportable on December 8, 1989; and, LER 89-011-00 was documented on January 5, 1990.
In review of NE0 2.25, 10 working days plus a maximum of 20 calendar days are prescribed from the originator to receipt at the unit superintendent for disposition.
The licensee staff informed the resident inspectors of numerous discussions on the
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reportability of the event between the Millstone 2 and corporate office staffs between September 6 - 27.
Plant incident report 89-93 dated September'6 was initially determined not reportable by the plant staff based on administrative control procedure 10.01 and emergency plan implementing procedures 4701-4. The three primary reasons for the determination that the event was not reportable were:
(1) the verification of safety injection actuation signal (SIAS)
components is identified in the emergency operating procedures; (2)
the probability of a loss of instrument air during the initial stages of a design basis accident (i.e. initiation of SIAS after a loss of instrument air pressure) is low; and, (3) the location of valve
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2-SW-3.2A was not in a " harsh" environment during a design basis event to reposition the valve on potential loss of instrument air.
All the above. explained reasons are considered mitigating factors for the event to assess safety significance, not conditions to determine reportability.
The inspector concluded that the general adequacy on timeliness of the reportability evaluation and notification to the NRC was insufficient. This is an unresolved item to determine licensee corrective actions to programmatically improve the reportability and NRC notification timeliness. (336/89-24-06)
In conclusion,-the NRC encourages and supports licensee initiatives
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for self-identification and correction of violations. The inspector reviewed 10 CFR 2 Appendix C subpart G " Exercise of Discretion" criteria; and noted that the incorrect air supply check valve
location resulted in a violation of the facility technical specifi-
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cations that was identified by the licensee (corporate engineer on September 6); the violation was considered to be of low safety consequence (a less significant violation of a technical specifica-
-tion-limiting condition for operation, not satisfied within the time
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allotment); it was reported (LER 89-011-00); it was corrected,
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including measures to prevent reoccurrence, and it was not a willful violation, or a violation expected to have been corrected by the licensee's corrective actions based on a previous-violation.
Thus, this is considered a non-cited violation.
(50-336/89-24-37)
8.4 Periodic Reporg Upon receipt, periodic reports _ submitted pursuant to technical specifNations were reviewed.
This review verified that the reported information was valid and included the required NRC data. The inspector was also ascertained whether any reported information should be classified as an abnormal occurrence. The following
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reports were reviewed:
Monthly Operating Report for November, 1989
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Monthly Operating Report for December, 1989
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No inadequacies were noted.
9.0 Management Meetings Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also discussed at the conclusion of the inspection. The licensee-identified
. violation and NRC additional concerns regarding inoperable containment radiation monitors was discussed and acknowledged by licensee management.
No proprietary information was covered within the scope of the inspection.
With the exception of TI 2515/104 survey data forms used to obtain fitness for duty training and program information, no written material was given to the licensee during the inspection period.
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